Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(5), p.051501_1 - 051501_10, 2020/10
Idomura, Yasuhiro; Ina, Takuya*; Ali, Y.*; Imamura, Toshiyuki*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.225 - 230, 2020/10
A new communication avoiding (CA) Krylov solver with a FP16 (half precision) preconditioner is developed for a semi-implicit finite difference solver in the Gyrokinetic Toroidal 5D full-f Eulerian code GT5D. In the solver, the bottleneck of global collective communication is resolved using a CA-Krylov subspace method, while the number of halo data communication is reduced by improving the convergence property using the FP16 preconditioner. The FP16 preconditioner is designed based on the physics properties of the operator and is implemented using the new support for FP16 SIMD operations on A64FX. The solver is ported on Fugaku (A64FX) and Summit (V100), which respectively show 63x and 29x speedups in socket performance compared to the conventional non-CA Krylov solver on JAEA-ICEX (Haswell).
Li, Y.; Hirota, Takatoshi*; Itabashi, Yu*; Yamamoto, Masato*; Kanto, Yasuhiro*; Suzuki, Masahide*; Miyamoto, Yuhei*
JAEA-Review 2020-011, 130 Pages, 2020/09
For the improvement of the structural integrity assessment methodology on reactor pressure vessels (RPVs), the probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed and improved in Japan Atomic Energy Agency based on the latest knowledge. The PASCAL code evaluates the failure probabilities and frequencies of Japanese RPVs under transient events such as pressure thermal shock considering neutron irradiation embrittlement. In order to confirm the reliability of the PASCAL as a domestic standard code and to promote the application of PFM on the domestic structural integrity assessments of RPVs, it is important to perform verification activities, and summarize the verification processes and results as a document. On the basis of these backgrounds, we established a working group, composed of experts on this field besides the developers, on the verification of the PASCAL module and the source program of PASCAL was released to the members of working group. This report summarizes the activities of the working group on the verification of PASCAL in FY2016 and FY2017.
Yamaguchi, Yoshihito; Hasegawa, Kunio; Li, Y.
Journal of Pressure Vessel Technology, 142(4), p.041507_1 - 041507_6, 2020/08
The phenomenon of crack closure is important in the prediction of fatigue crack growth. Several experimental data indicate the closing of fatigue cracks both under negative and positive loads at constant amplitude loading cycles, depending on the magnitude of stress amplitude and stress ratio. Appendix A-4300 of the ASME Code Section XI provides two equations of fatigue crack growth rates expressed by the stress intensity factor range for ferritic steels under negative stress ratio. The boundary of two fatigue crack growth rates is classified with the magnitude of applied stress intensity factor range, in consideration of the crack closure. The boundary value provided by the ASME Code Section XI is validated in this study through an investigation of the influence of the magnitude of the applied stress intensity factor range on crack closure, with the application of fatigue crack growth tests using ferritic steel specimens in air environment at room and high temperatures. Crack closures are obtained as a parameter of stress ratio, and herein, were found to occur at a smaller applied stress intensity factor range, as opposed to the definition given by Appendix A-4300.
Okuda, Yukihiko; Kang, Z.; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
In case of projectile impact to reactor building of nuclear power plants, stress waves due to the projectile impact propagate from the impacted wall to the interior of the structure. It is an important issue to assess the dynamic response generated with projectile impact for safety related internal equipment because stress waves are likely to excite high-frequency vibrations of internal equipment in the reactor building. The OECD (Organization for Economic Co-operation and Development) / NEA (Nuclear Energy Agency) launched the IRIS (Improving Robustness Assessment Methodologies for Structures Impacted by Projectiles) benchmark project in order to assess the dynamic response for nuclear facility by projectile impact and the third phase of IRIS (IRIS 3) contributes to the investigation on the dynamic response of reinforced concrete (RC) structure with internal equipment. We have participated in the IRIS 3 and have performed the calibration analysis for projectile impact test on the structure which models a reactor building and internal equipment. Specially, we have developed and validated a numerical approach to investigate impact response of the RC structure with internal equipment through the calibration correction. This paper presents partial simulation results from dynamic response of the RC structure with internal equipment and discusses the effect of supporting condition of the internal equipment and stress wave propagation.
Nishida, Akemi; Kang, Z.; Okuda, Yukihiko; Tsubota, Haruji; Li, Y.
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08
Studies on the local damage to reinforced concrete (RC) panels subjected to projectile impact have mainly focused on collisions that occur at an angle normal to the structure; thus, research on oblique impact is scarce. Therefore, we conducted research focusing on oblique impact to enable more realistic impact assessment of projectile collisions. To date, the validity of the analytical method has been confirmed by comparing the results with those of previous tests, and the local damage of RC panels that have collided with projectiles has been analytically investigated focusing on the impact angle. Therefore, this study aims to confirm the validity of the analysis method by conducting impact tests under various conditions including the impact angle, and obtaining data for validation. This paper outlines the test for the local damage of RC panels subjected to normal and oblique impact.
Plompen, A. J. M.*; Cabellos, O.*; De Saint Jean, C.*; Fleming, M.*; Algora, A.*; Angelone, M.*; Archier, P.*; Bauge, E.*; Bersillon, O.*; Blokhin, A.*; et al.
European Physical Journal A, 56(7), p.181_1 - 181_108, 2020/07
The Joint Evaluated Fission and Fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides U, U and Pu, on Am and Na, Ni, Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yileds, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 is excellent for a wide range of nuclear technology applications, in particular nuclear energy.
Onodera, Naoyuki; Idomura, Yasuhiro; Ali, Y.*; Shimokawabe, Takashi*; Aoki, Takayuki*
Dai-25-Kai Nippon Keisan Kogaku Koenkai Rombunshu (CD-ROM), 4 Pages, 2020/06
We have developed the stencil-based CFD code JUPITER for simulating three-dimensional multiphase flows. A GPU-accelerated Poisson solver based on the preconditioned conjugate gradient (P-CG) method with a multigrid preconditioner was developed for the JUPITER with block-structured AMR mesh. All Poisson kernels were implemented using CUDA, and the GPU kernel function is well tuned to achieve high performance on GPU supercomputers. The developed multigrid solver shows good convergence of about 1/7 compared with the original P-CG method, and 3 speed up is achieved with strong scaling test from 8 to 216 GPUs on TSUBAME 3.0.
Hasegawa, Kunio; Li, Y.; Lacroix, V.*; Mares, V.*
Journal of Pressure Vessel Technology, 142(3), p.031506_1 - 031506_7, 2020/06
Bending stress at plastic collapse for a circumferentially cracked pipe is predicted by limit load criterion provided by the Appendix C of the ASME Code Section XI. The equation of the Appendix C is applicable for pipes with both external and internal surface cracks. On the other hand, the authors have developed a more precise equation. From the comparison of Appendix C equation and the new equation, the plastic collapse stress estimated by the Appendix C equation gives less conservative bending capacity prediction for external cracked pipes with thick wall thickness and large crack angle. This paper discusses the limitation scope to use the limit load criterion of the Appendix C equation.
Mano, Akihiro; Katsuyama, Jinya; Li, Y.
Mechanical Engineering Journal (Internet), 7(3), p.19-00567_1 - 19-00567_11, 2020/06
Non-destructive examinations (NDEs) have an important role in assurance of the structural integrity of nuclear components including pipe lines. In Japanese nuclear power plants, NDEs are performed for welds in piping in accordance with the rules such as the Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers. For the welds where stress corrosion cracking (SCC) is not postulated, NDEs are performed in each 10-year interval. For each interval, the extent of examination is specified in the rules. In general, there are two kinds of sampling method for selecting welds to be examined in each interval considering the specified extent of examination. The first method is the fixed location sampling method, in which welds for NDEs are same as those examined in the last interval. The second method is the random location sampling method, in which welds for NDEs are selected from those not examined in the last interval. The selection of the sampling method is important to assure the structural integrity of piping. Probabilistic fracture mechanics (PFM) analysis which is one of rational structural integrity assessment methods can quantitatively calculate failure probability of welds in piping considering aging degradation mechanisms such as SCC and fatigue as well as crack detections and repair of cracked welds through NDE. In this study, to clarify the influence of the sampling methods on structural integrity of piping, we evaluated the failure probability of a typical nuclear piping considering NDEs based on the two sampling methods through PFM analysis. From the results, we clarified the quantitative influence of two sampling methods on failure probability of piping.
Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*
Mechanical Engineering Journal (Internet), 7(3), p.19-00573_1 - 19-00573_14, 2020/06
Kang, Z.; Nishida, Akemi; Okuda, Yukihiko; Tsubota, Haruji; Li, Y.
Mechanical Engineering Journal (Internet), 7(3), p.19-00566_1 - 19-00566_20, 2020/06
Most impact research has been presented on the basis of impact tests and numerical analysis performed by rigid projectile impact perpendicular to the target structure. On the other hand, there are only few reports on impacts at an oblique angle. To evaluate more realistic conditions regarding issues related to oblique impacts to reinforced concrete (RC) structures, we have proposed an analytical method to estimate the local damage to RC structures by an oblique impact and have validated the evaluation approach by comparison with experimental results. At present, we have finalized simulation analyses of oblique impact assessments on RC panels using rigid/soft projectiles with a flat nose shape utilizing the validated approach. Furthermore, in this study, we focus on impacts caused by rigid/soft projectiles with a hemispherical nose shape. The same analytical method is applied to simulate the structural damage caused by an RC panel due to a rigid/soft projectile with a hemispherical nose shape. Results on the penetration depth of the RC structure and the energy-contribution ratio are presented. By comparing the results of local damage to an RC structure caused by projectiles with flat and hemispherical nose shapes, the influence of the nose shape of the projectile on local damage of the RC panel has been investigated.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Osaka, Masahiko; Li, Y.
Mechanical Engineering Journal (Internet), 7(3), p.19-00560_1 - 19-00560_12, 2020/06
Since the Fukushima Daiichi nuclear power plant accident, we have been developing a failure evaluation method that considers creep damage mechanisms using detailed three-dimensional finite element analysis model of lower head including penetration, stub tubes, and weld parts, etc., for the early completion of the decommissioning of the nuclear power plants in Fukushima Daiichi. For the finite element analysis, we have been obtaining material properties for which no data are provided in existing databases or in the literature. In particular, creep data corresponding to the high temperature region near the melting point of materials is important in evaluating creep deformation under severe accident conditions. In this study, we obtained the uniaxial tensile and creep properties for low-alloy steel, stainless steel, and Ni-based alloy. In particular, creep test data with long rupture times at high temperatures are expanded using a tensile test machine that can measure the elongation of test specimens in a noncontact measurement system. The parameters related to the failure evaluation were improved on the basis of the expanded creep database.
Bao, S.*; Cai, Z.*; Si, W.*; Wang, W.*; Wang, X.*; Shangguan, Y.*; Ma, Z.*; Dong, Z.-Y.*; Kajimoto, Ryoichi; Ikeuchi, Kazuhiko*; et al.
Physical Review B, 101(21), p.214419_1 - 214419_8, 2020/06
Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.; Onizawa, Kunio
Journal of Pressure Vessel Technology, 142(2), p.021906_1 - 021906_11, 2020/04
Azuma, Kisaburo*; Li, Y.; Xu, S.*
Journal of Pressure Vessel Technology, 142(2), p.021207_1 - 021207_10, 2020/04
Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04
Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei*; Li, Y.; Yoshimura, Shinobu*
Journal of Pressure Vessel Technology, 142(2), p.021205_1 - 021205_10, 2020/04
no abstracts in English
Lokotko, T.*; Leblond, S.*; Lee, J.*; Doornenbal, P.*; Obertelli, A.*; Poves, A.*; Nowacki, F.*; Ogata, Kazuyuki*; Yoshida, Kazuki; Authelet, G.*; et al.
Physical Review C, 101(3), p.034314_1 - 034314_7, 2020/03
The structures of the neutron-rich Co isotopes were investigated via () knockout reactions at the Radioactive Isotope Beam Factory, RIKEN. Level schemes were reconstructed using the coincidence technique, with tentative spin-parity assignments based on the measured inclusive and exclusive cross sections. Comparison with shell-model calculations suggests coexistence of spherical and deformed shapes at low excitation energies in the Co isotopes.
Sun, Y. L.*; Obertelli, A.*; Doornenbal, P.*; Barbieri, C.*; Chazono, Yoshiki*; Duguet, T.*; Liu, H. N.*; Navrtil, P.*; Nowacki, F.*; Ogata, Kazuyuki*; et al.
Physics Letters B, 802, p.135215_1 - 135215_7, 2020/03
no abstracts in English